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Monte Carlo Calculations Using MCNP4B for an Optimal Shielding Design of a 14-MeV Neutron Source


The shielding for 14-MeV neutrons from a D-T generator at SLAC needs to be optimized in terms of shield thickness and cost. Shielding calculations were made using the MCNP4B Monte Carlo code and the ENDF/B-VI cross section set. A few materials were studied: iron, borated polyethylene, three types of normal and heavy concrete, and mixtures of iron and borated polyethylene. The effects of shielding geometry (sphere, cube, and the proposed geometry) were also examined. The transmission results of ambient dose equivalents, as well as the attenuation lengths and the average energies, for neutrons and gammas outside the shielding are presented.

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